The present invention relates to a method for purifying actinides present in a low oxidation state in an aqueous solution containing fission products, the actinides being from the group of uranium (IV), neptunium (IV) and plutonium (III), by extracting the actinides from the aqueous solution while keeping the fission products in the aqueous solution.
A method for purifying plutonium in 3 valence state is used, for example, in reprocessing processes of the PUREX type for spent nuclear fuel and/or breeder materials. To better clarify the multitude of sometimes complicated process steps in such a process, it is the custom to combine several process steps to so-called cycles at the ends of which the materials, such as uranium or plutonium, for example, are present in aqueous solution. The heart of the PUREX process, for example, after dissolving the spent nuclear fuel substance in an aqueous solution of nitric acid, comprises two extraction cycles. The first extraction cycle is a common extraction cycle in which uranium and plutonium are simultaneously separated from the aqueous solution containing the dissolved nuclear fuel and fission products into an organic extraction medium, such as, for example, tributyl phosphate (TBP). During this extraction there is a rough separation of the fission products which remain in the aqueous solution from the uranium and plutonium. The uranium and plutonium in the organic extract can then be transferred to separate aqueous stripping solutions and a second or fine extraction cycle for each of the uranium and the plutonium aqueous solutions can then be made with another organic extraction medium, which again can comprise TBP, to further purify these elements. Two extraction cycles utilizing, for example, tributyl phosphate (TBP) as the extraction agent are not sufficient, however, to produce high degrees of purity for either the uranium or the plutonium, and particularly, are not sufficient with respect to removing the fission products zirconium, niobium and ruthenium. Furthermore, with only two extraction cycles, the entire process is narrowly limited, particularly with respect to the extraction conditions or to the apparatus involved. For this reason, an additional or third solvent extraction cycle is sometimes introduced before the fine purification of the actinides to produce a further purification for uranium and plutonium. This third extraction cycle comprises separate extraction treatments on the two aqueous stripping streams which contain the uranium and plutonium, respectively.
It has also been proposed to use, instead of the two separate third extraction cycles for U and for Pu, a codecontamination cycle which takes place before the common extraction cycle and which would have the advantage of requiring very little in apparatus. It has also been proposed to us an anion exchange process or a sorption process instead of a third solvent extraction cycle to bring about further purification.
As stated above, in the common extraction cycle for separating the uranium and plutonium from the fission products, the nitric acid solution from the fuel suspension which contains the actinides and fission products is initially treated in an extraction process in which uranium and plutonium are transferred to the organic extraction agent and the majority of the fission products remain in the aqueous solution. The plutonium in the nitric acid solution has a valence state of 4 and passes into the organic phase with a valence state of 4. The organic phase is separated, scrubbed, and subjected to a re-extraction treatment selective for plutonium and in which uranium remains in the organic phase and the plutonium is separated from the uranium. This re-extraction treatment for selective extraction of plutonium, for example, can comprise an electrolytic reduction of the plutonium in the organic phase from the 4 valence state to a 3 valence state, and whereupon the Pu (III) can then be re-extracted to an aqueous phase with the aid of an aqueous hydrazine nitrate or hydroxyl ammonium nitrate solution. The aqueous phase containing the Pu (III) then is washed with an organic extraction agent solution to extract small quantities of uranium which has passed into the aqueous solution. The uranium is then re-extracted from the organic phase into an aqueous phase with diluted nitric acid.
Depending on whether a codecontamination cycle is introduced before the common U-Pu extraction and separation cycle, the plutonium in the prior art recovery processes after passing into the aqueous phase as Pu (III) from the organic extract, undergoes one further extraction cycle, if codecontamination was used or undergoes two further extraction cycles if codecontamination was not used, in order to further purify it and separate fission products and uranium which have been carried along and not previously removed. The known process for purifying the Pu (III) in the aqueous phase begins, after the above-described wash of the aqueous phase, by oxidizing the Pu (III) in the aqueous phase to Pu (IV). After oxidation, the Pu (IV) is extracted from the aqueous phase to an organic phase by using a TBP extractant. A considerable portion of the carried-along fission products, particularly zirconium and ruthenium, again remain in the aqueous phase. Thereafter, the organic phase containing the Pu (IV) is separated from the aqueous phase and washed. After separation and washing of the organic phase, the plutonium in the organic phase is again reduced to Pu (III) by means of electrolytic reduction and is re-extracted into an aqueous phase with a hydrazine nitrate solution or hydroxyl ammonium nitrate solution to complete the first Pu (III) purification cycle. In case no codecontamination cycle was used before the common U-Pu extraction cycle, a further Pu (III) purification cycle is performed by repeating the just-described steps.
The known processes for treating the spent nuclear fuels, as described above, have a number of drawbacks. For example, almost all of the valuable neptunium goes into the aqueous waste solutions together with the fission products during the various cycles. Loss of the neptunium occurs particularly during the common U-Pu extraction cycle, but also occurs during the purification cycles. Further, the systems are too complicated with respect to space requirements and expenses.
In addition, the process step of oxidizing the Pu (III) to Pu (IV) in the plutonium purification process results in further drawbacks. The previously employed oxidation of Pu (III) to Pu (IV) is effected by an addition of sodium nitrite (NaNO.sub.2) or by the introduction of nitrogen dioxide (NO.sub.2) or dinitrogen tetroxide (N.sub.2 O.sub.4) to the aqueous solution containing the Pu (III) from the electrolytic reduction. Hydrazine nitrate is still present in this aqueous solution and is destroyed during the oxidation step, for example, according to the formula: EQU N.sub.2 O.sub.4 + H.sub.2 O .revreaction. HNO.sub.3 + HNO.sub.2 EQU n.sub.2 h.sub.5.sup.+ + hno.sub.2 .fwdarw. hn.sub.3 + 2h.sub.2 o + h.sup.+
or EQU N.sub.2 H.sub.4 + 2HNO.sub.2 .fwdarw. N.sub.2 + N.sub.2 O + 3H.sub.2 O
when N.sub.2 O.sub.4 is used, the oxidation of Pu (III) takes place according to the following formula EQU Pu.sup.+.sup.3 + N.sub.2 O.sub.4 .fwdarw. Pu.sup.+.sup.4 + NO.sub.2 + NO.sub.2.sup.-
in this prior art oxidation process, high excess quantities of NaNO.sub.2 or N.sub.2 O.sub.4 are required, for example, about ten times the stoichiometric quantity. The apparatus required for high throughputs is also considerable. The gas must be introduced in giant columns and the excess of oxidation agent must be removed in a subsequent stripping column by blowing in air. This leads to large quantities of contaminated exhaust gases. The costs for such a process are high. For example, the costs of the N.sub.2 O.sub.4 alone can amount to about 200 to 250 dollars per ton of light water reactor fuel to be reprocessed.
It has been reported, by D. Cohen, JOURNAL OF INORGANIC AND NUCLEAR CHEMISTRY, Vol. 18, 1961, pages 207 to 210, that, in the course of tests to electrochemically display four valence states of plutonium in perchloric acid solutions, the electrolytic technique produces pure solutions of a given oxidation state and is particularly useful for the production of relatively concentrated (0.02 M) solutions of Pu (V). The recorded current-voltage curves in these tests indicate that Pu (III) can be oxidized to Pu (IV) in HClO.sub.4 solution at a potential between -1.3 and -1.4 volt. The tests used to produce the current-voltage curves were conducted at 25.degree. C. and exclusively in 1 M HClO.sub.4 solution. The literature does not show, however, that anodic oxidation of Pu (III) would be possible in a nonuniform aqueous nitric acid solution which is distinguished by the number and variety of its contents and their reactions, such as the aqueous nitric acid solutions that are found during the reprocessing of nuclear fuels and which contain various fission products and other components.